Nuclear reactors using a primary coolant of the core of the reactor, such as nuclear reactors cooled by pressurized water, comprise a vessel in which is disposed the core of the reactor constituted by fuel rods and a primary cooling system comprising at least one loop in which is disposed a steam generator inside which the primary coolant of the reactor provides the heating and the vaporization of the feedwater. Each of the loops of the primary circuit comprises pipes of large diameter and thickness in which the primary coolant flows.
One of the pipes, called the hot leg, connects the vessel to the steam generator and provides the transfer of the heated primary coolant in contact with the core to the primary part of the steam generator.
Another pipe, called the cold leg, provides the return of the coolant to the vessel, after its passage through the steam generator.
In order to provide the control and the operational monitoring of the nuclear reactors, it is necessary to measure the temperature of the primary coolant which absolutely has to be maintained within a predetermined range, in order to provide a satisfactory operation of the nuclear reactor.
There are known devices for measuring the temperature of the primary coolant of a nuclear reactor, comprising a metallic body called a scoop fixed in the wall of a reactor coolant pipe, in such a manner that a portion projects inside the reactor coolant pipe in the form of a thimble and pierced by at least one channel communicating with the inner volume of the reactor coolant pipe by a plurality of openings distributed in a substantially radial direction in relation to the pipe. A temperature measurement probe is fixed on the inside of the channel of the scoop, in such a manner as to come into contact with the primary coolant penetrating into the channel of the scoop by the openings traversing the wall of the thimble.
It may also be necessary to provide an opening for outflow of the coolant traversing the wall of the thimble in a zone situated opposite the openings for inflow of the coolant.
These devices make it possible to carry out direct measurement of the temperature of the primary coolant, taking into account the stratification existing in the primary coolant pipe and manifested by a temperature gradient in the radial direction, owing to the distribution of openings for inflow of the coolant into the scoop.
However, the temperature measurement obtained is not entirely reliable and the response time of the device is not extremely rapid, due to the fact that the coolant is not mixed and homogenized efficiently before the measurement and that its speed of circulation inside the channel of the scoop is relatively low.
There is also known, from French Patent Application 90 16493, filed Dec. 28, 1990 by the present applicant, a device for measuring the temperature of the primary coolant of a pressurized water nuclear reactor which comprises at least three elements for sampling coolant traversing the wall of a substantially horizontal portion of the hot leg, which elements are distributed on the periphery of the hot leg, which element is connected by a pipe outside the hot leg, to each of the sampling elements.
A temperature probe is disposed on the inside of each of the sampling elements. Moreover, a temperature probe may be also disposed inside the element for reintroducing the fluid.
This device makes it possible to carry out a good mixing of the primary coolant before the measurement, but the flow rate of the primary coolant in the measurement circuit is relatively low, so that the response time of the measuring device is not extremely rapid.
Furthermore, the device has a relatively complex structure and comprises pipes on the outside of the reactor coolant pipe. Difficulties may arise thereby, in particular when a thermal insulation material is placed around the reactor coolant pipe.